ML20126H591
| ML20126H591 | |
| Person / Time | |
|---|---|
| Site: | Zion File:ZionSolutions icon.png |
| Issue date: | 04/07/1981 |
| From: | Varga S Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20126H588 | List: |
| References | |
| NUDOCS 8104160079 | |
| Download: ML20126H591 (59) | |
Text
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, , ' .o UNITED ::TATEs
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COMMONWEALTH EDISON COMPANY DOCKET NO. 50-295 ZION STATION UNIT 1 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No.62 License No. DPR-39
. The Nuclear Regulatory Commission (the Commission) has found that:
A. The applications for amendment by Commonwealth Edison Company (the licensee) dated April 22,1930, May 30,1980, and October 3, 1980, comply with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act) and the Commission's rules and regulations set forth in 10 CFR Chapter I;
- 3. The facility will operate in conformity with the applications, the provisions of the Act, and the rules and reculations of the Commmssion; C. There is reasonable assurance (i) that the activities authorized ~
by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be ccnducted in compliance with the Commission's regulations:
D. The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and E. The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied.
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- 2. Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attacnment to this license amendment, and paragrar.h 2.C.(2) of Facility Operating License No. CPR-39 is hereby amended to read as folicws:
(2) Technical Scecifications The Tecnnical Scecificaticnt contained in Apcendices A and 3, as revised thrcugh Amendment No. 62' , are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical Specifications.
- 3. This license amendment is effecti'.e as of the date of its issuance.
FOR THE NUCLEAR ',ESULATORY COMMISSIO!i
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.StevenA.f(hbh
!arga, Chie Operating Reac:crs Br,ench #1 Division of Licensingd At achment: -
nangas to the Technical 5;ecificatiens .
Cate of :ssuance: April 7,1981
$2 2tc 4 A UNITED STATES ' .
y $p e,.gy NUCLEAR REGULATORY COMMISSION
- 4.Og, j C WASHINGTON. D. C. 20555
% he COMMQ A.'EALTH EDISON COMPANY, DOCKET NO. 50 304 ZION STATION UNIT 2 AMENDMENT TO FACILITY OPERATINF. LICENSE Amendment No. 59 License No. DPR 48
- 1. The Nuclear Regulatory Commission (the Commission) has found that:
A. The applications for amendment by Commonwealth Edison Company (the licensee) dated April 22,1980, May 30,1980, and October 3, 1950, comply with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act) and the Commission's rules and regulations set forth in 10 CFR Chapter I; B. The facility will operate in conformity with the applications the provisions of the Act, and the rules and regulations 6f the Commmssion; C. There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health -
and safety of *Se public, and (ii) that such activities will be
. conducted in ccmpliance with the Commission's regulations:
D. The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and
! E. The issuance of this amendment is in accordance with 10 CFR Part
! 51 of the Commission's reculations and all applicable requirements have been satisfied.
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' 2. Accordingly. the license is amended.by changes toithe Technical ;[
. Specifications - as indicated in~ the ' attachment to this ' license !
- ' amendment, and paragraph 2.C.(2) of Facility. Operating License -
No. DPR 48 is hereby amended to read as follcws:-
~
(2) Technical Scecifications
- The Technical'Specificaticns contained in Appendices 1
-A and B, as revised through Amendment No. 59 , are .
t hereby ~ incorporated in the. licensc. The licensee shall operate .the facility in acecrdance with the Technical '
Specifications. i
- 3. This license amendment is effective as of :ne date of its issuance. i FOR( HE NUCLfAR REGULATORY COMMISS
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- Steven A. Yarga, unie6
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' Operating Reac:crs Bra (nch #1 Divisien;of Licansing - r i
Attachment:
Changas to the Technical Specifications .
- ate of Issuance: April 7,1981 1
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. ATTACHMENT TO LICENSE AMENDMENTS AMENDME!;f NO. 62-TO FACILITY OPERATING LICENSE NO. DPR-39 AMENDMENT NO. 59 TO' FACILITY OPERATING LICENSE NO. DPR-48 DOCKET NOS. 50-295 AND 50-304 Revise Appendix A as follows:
Remove Pages Insert Pages iii iii vii vii 6 6 12 12 19 19 24 24 30 30-31 31 32 32 ,
36 36 40 40 41 41 ,
42 42 44 44 46b 46b c
47b 47b 48 48 49 49 50 50 51 51 52 52 54 54 55 55 68 68 _
6Ba 68a 69a 69a 70 70
.. ~. A.
F ATTACHMENT TO LICENSE AMENDMENTS (C.ONT. )
I Remove Pages Insert Pages 71 71' l 72 72 !
74g 74g :
79 79 !
80 80 ;
i 81 81- l 95 95 96 96 105 105
. 119 119 156 156 173A 173A-187 187 297 297..
300 300 301 301 [
302 302 302A 302A 303 303 308 308 329 329 330 330 331 331 305 305 306 306 314 314 I
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4 Tal l e o f Con t enits- (Con t. i nised) .
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" SIIIWE T I.T.Attri:
- l. I,"I *rI'!G rotinI'i'l o'i rop OPI:9 ATI o'1 n rnt'I oPH PIIT PAGF I 1,7 rtcan Cetierator rmeroency !'ea t 9enoval 1.7 156 Stein I.ine Tafety Valves 4.7.1 150-t.7.I d.7.2 159 3'
I5.'.? Auxiliarv Pee 1 eater Puro Systen n ases l
-1.9 I:rer<tency core coolina and core coolinn Sunnort d.P 164 1.a.) Cr-n tr i f una l charni no Pusan Fust.em A.O.1 164 1.a.1 ra'ety Iniection Punp System a.o.? 169 1.c.1 Pesidual l'ea t nenoval Punp Syst.cr- d.p .3 17e
- 1. a . .! F i ster testinn of Centri fueal clearni nc, Fafety Iniection, and nesidual l'en t nenoval Purn Pt' stem's l 4.3.4 l 171 1.a.4 Acetimitt a tor Fys t.cm A.P.5 174
, 1.9.r Corponent Coolino Svstem d.R.6 175 Service tfater System A,8,7 17 sg ;
1.9.7 I!ydronen Control Systems 4.R.8 1RO i 1.9.9 '
1.9.9 1:cu ipmen t for Fvaluatino Post i.oCA d.R.9 1RJ Pases I 3.a Containrent Isolation Systens 4.0 147 3.".! Isniation Valve Seal Fater System d.9.1 197
- 1.a.? Penetration Pressurization Systens 4.4.2 loe Containment Isolarion Valves 4.9.3 199
- j' 1.'.1 t."..' "ain Stean Tsolation Valves and Pypasses d.".4 2 n r.
- 1. * . i Con '. a i niaen t Irrecritv 4.a.5 201 .
- Ha9es 7
- t. ! o con ta i nrent S t ructura l Intenrity 4.10 217
- 1. t '$ .1 Portainrent I. eat: ace P.z:t.e Testi ng 2.19.1 212 2.la.2 Conta i nnent Tendon '"esti ng A.in.2 71s En i Anctiorasyes and Adjacent Concrete .1urf aces
~
1.14.3 T ospec ti on 4.*10.3 217 3.14.4 Cortainment T.iner Inspection d.10.4 719-1.19.5 Containrent Pressure 4.10.5 219
- 3.In.F Cortainment Ternerature A.]".6 219 "ases 1.11 n a?iocict ive f.icuids /. . ! ) '??
rases -
1. " Rinactive cases 3.]? 21n rases 1
Amendnent ilos. 621. 59 111
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4 LIST OF FIGURES (Continued)
Figure Page 3.3.2-1 Reactor Coolant System Heatup Limitations 84 3.3.2-2 Reactor Coolant System Cooldown Limitations 85 3.3.2-3 Ef fect of Fludice and Copper Content on Shif t of 86 RT for Reactor Vessel Steels Exposed to 550 CF
- Temperature 3.3.2-4 Fluence at 1/4T and 3/4T as a Function of Full Power 87.
Service Years-3.4-1 High Steam Line Flow Setpoint 131a 4.16-1 Location of Fixed Environmental Radiological 278 Monitoring Stations Commonwealth Edison Corporate and Station Organization 329 6.1.1 Zion Shift Hanning Chart 331 6.1.2 "iI Amendme.:* Nos. 62 & 59
2
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L G. Quadrant Power Tilt latio K. Operable i The tiuadrant power tilt ratio is Properly installed in the system and defined as the ratio of the maximum upper capable of performing the intended excore detector current to the functions in the intended manner as average of the upper excore detector verified by testing and tested'at
, currente or the ratio of the maximum the frequency. required by the lower' excore detector current to the Technical Specifications. i average of the lower excore detector currents whichever is greater. L. Operating Performing the intended functions in the intended manner.
f t. Operating Cycle
- 11. Hated Thermal Power The interval between the end of:one A s teady-s ta te reactor core output - of maior refueling outage and the end of '
3250 MWg per unit. the next subsequent major refueling l outage per unit.
- 1. Heactor Pressure
, N. Surveillance' Interval The pressure in the steam space of
. a pressurizer. Each Surveillance Requirement shall be 1
performed within the specified time interval J. Refueling Outage with:
When Refueling Outage is use'd to g a. A maximum' allowable extension not
{ designate a surveillance interval per unit, to exceed 25% of the surveillance the surveillance will be performed interval; and during the refueling outage or up to six months before the refueling b. A total maximum combined intervol
- outage. When a refueling outage time for any three consecutive occurs within 8 months of the surveillance intervals.not to' previous refueling outage for a unit, the exceed 3.25 times the specified surveillance testing need not be interval.
performed. .The maximum interval
- between surve113 nnee tests is
- 20 months per unit. ,
Amendment Nos. 62 & 59 6
. . . . ~ _ _ _ . . _ . . _ .- . _ - _ _ _ _. - _ . . _ . _ . _ __. . .-~ -
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7 SAFETY LIMIT LIMITING SAFETY SYSTEM SETTING
- 7. Low reactor coolant pump motor frequency:
2: 57.0 cps.
- 8. Undervoltage to reactor coolant pump motors: h 70% of normal.
C. Other reactor trips
- 1. Iligh pressurizer water level: $E 92% of
, span.
- 2. Low-low steam generator water level:
2L 10% of narrow range instrument.
- 3. Steam feedwater flow mismatch: fi 60% of nominal 100% steam flow rate in coincidence I with low steam generator water level --
2 10% of narrow range instrument span.
- 4. Safety Injection - Trip settings for safety injection are detailed in Section 3.4.
- 5. . Turbine Trip
- 6. Power range, positive high neutron-flux rate SE 15% of rated flux in 5 sec.
- 7. negative high neutron flux Powerrange,loftheratedfluxin5sec.
rate $il-15%
- 8. Manual reactor trip..
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12 Amendment Nos. 62 & 59
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- ie rve rn a r.- base.1 on t.he follovin<;
J :in. - l . . i r he et- ( 'ia n ne ! 'a e -l.o r n : (2) 4 i
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= 1 . '3% {1 + 0.2 (1-P)]
4 t.dere P in the fraction of rated power, The hot channel Factors are also and on a utiti analysis as described,in fM 6e 1 to a w o m for j lief erence 3. The e:tpression for r)gg um hme h* mlmsihim M dm full-length control rods that is or 9 < n < 1. ') renresents the e*Fect r
alloecd before the reactor trin set i* oi. ra.!ill r>over shanen of t.he control noints are reduced and rod 'eithdraval i . 's 11. the innertion limits. Islock and load runbach nav be required.
] (5) 11od withdrawal block and load runback
- occurs if reactor trin set noints are
! apnroached within a fixed limit. (6)
The Reactor control anel - Protection -
S** stem is desioned to oretrent any anticinated combination'of transient.
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1 l (2) PSArt Annendix 3A 9ection 5.3 (3) PFAR Anpendix 3^ 9ection 4.3 (5) P. CAP. Section 14.1.3 (6) PSAR Section 7.2.2 i
- Amendnent flos. 62 & 59 19 i
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Ilo . Of . Min ismism Minimum Operator Action
. Cliatuiels Operalile l>egsee of If Colusun 3 or.4 i lae.ic t or Ts ip Ch.ume t tio. Of Setpointet -
Isescri L>t ion Per lisil t Cliasme l s To Trit Cleannels t e t Itedundancy t t e Can Hot lie Het +
- 1. FLime.n l iteactor Trip 2 l- 1 0 Maintain Ilot Sliutdown** N'. A .
- 2. Powes Ita ngt' liigli l'linx . ,
25% of stated ticut ron .
Ilow nel poltet )-intei lockeil willi P-lu 4 2 3 2 liaint'ain Ilot Slmtinown*
- Flux 109% of Itated
- 1. l'e se r Itange liigli Fleix (higli set poisell 4 2 3 2 Maintain llot Sliutdown** taeutroni Flux Power it.inge tilgli Flux Hal e 4 2 _I 2 Maintain llot Sliutdown** .15% of Itated ticult on
- 4. Flux /5 sec.
15% of . Itated Heint ron
- 5. Ib'gative Power Itange Fleax it.i t.c 4 2 3 2 Mciintain Ilot Slautdown** l' lux /5 sec.
E. . Suu re e Itange Heist roni Flux- 3 2 1 1 0 Maintaisi llot Shutduwu S** 10 counts /sec.
Interlocked with P-lo and P-6 for CSD if that condition exists) 251 uf Itated fleut ron '
. 1. Intes snediat e llan.ge tscution 0 Haintain 1500 Sluttdowra* . Flux l' A ux-Isiter lockawl wi t h P-10 2 1 1 4 Ioops 4 2 3 2 Maintain Ilot St.utdown*
- Actuaia T?.Programn.e'd H. Overtemperatuse a T, overtemperature 6 ? I lingen 1 2 3 1 Maintain flot Shistdown*
- Setpoint
- 9. uvettiower 6 T, 4 loops 4 2 3 2 Maintain llot Shutdown ** Ac tua l A T 2 Pa sw3 rammed overpower a T, 3 loops -3 2 3 i Maintain flot Slautdown*
- Setpoint i
i l ei . Psennurizer IAna Pressune - Malutaisillot Shutdown ** 1825 psly intes lockeil witti P-7 4 2 3* 2 II. Presnur izer liigli l'resnus e 4 2 3 2 Heilntair tint Slintdown*
- 2185 psig
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TAlli I' 3.1-1 Itoaa tor Pr ot ectiori Systein - I.imi li en) Operation Cunditions and Setpoints 30 Amendment Hos. 62 & 59 .
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j l ett er lochest wi t h 4.. P-1 1 per l e wign 2 gier Irw*p 2 gw'r lewsp I flaisitalie Itot Sinat elnwn* *
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- 1.. PvP n.m usi...rie.Piency -
1 inst er locke.I .*ill P- f anet P-83 I per lass 2 1 2 flaisitain tint Sinst elnwn * * $7.o sia [
t I f. . prt.Itseaker Tsip -
Int ce law keel witis:
- a. P-7, 4 l e wigm 4 liscakern 2 1 2 Maintain Isot Slnat elown*
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l le. P-8, 4 lewsp s 4 hecakern I 1 2 ' Plai nt ales lint Shut.Inwn*
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- 17. lesw nicam Genesator I. eve l l es 2 pe r Irmp i por lamp 1 0 Maint ain plot Shut elown*
- 25% of narrow l Icvel caevic 'spase -
z cu linet i.leew-c w i t ti Tee.1 f i nw -
a nicam Tlow minmatch 2 per l amp i lies l ong, 1 O Plaintain Ilot Shutelown*
- anel 0.7 x I tt
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i 10 lesw-lere Si cam Generat nr 1 per inop 2 gw r Innp , 2 per lone i Maintain flot Shuttinwn *
- Int of narrrw
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i 1
0' O O V LIMITI!iG CONDITION FOR OPERATION' SURVEILLANCE REQUIRE!*.ENT N.
i.
4.2.1.B once a shift while remaining in this_ con- ;
! dition. During heatup, the boron concen- s i tration in the reactor. coolant loops and l pressurizer shall be sampled every 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.
The reactor coolant loop boron concentra- ,
tion must not decrease by more than 50 ppa. d between successive 4 hour4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> samples. The 1-pressurizer boron concentration must not
!.: be more than 200 ppm less than the' reactor i coolant loop boron concentration.
j C. Unit Startup C. Startup i . .
. 1. Immediately prior to startup, the reactor 1. The Tavg of each reactor coolant loop I coolant temperature shall be shown to be shall be logged before attempting to greater than the temperature above which bring a reactor: critical. ~,
l the moderator temperature coefficient is l always negative (except during low power j l physics tests) and greater than 500 'F. '
i
- 2. When a reactor is approaching criticality, 2. Not Applicable.-
{ the shutdown banks shall be fully with-l drawn in sequence (shutdown bank A,B,C,D)
! before any other rods are withdrawn. The .
I control group rods shall be no further in-l serted than the limits shown byl Figure
, 3.2-2 for Unit 1 and Figure 3.2-4 for i Unit 2 for 4-loop operation and Figure
- 3.2-3 for Unit =1 and Figurs.3.2-5 for '
j Unit 2 for 3-loop operation when critical- ;
- ity is attained.
i
} -D. Power Operation D. . Power Operation t 1. -When a reactor is critical,. except for 1. Rod operation shall be verified:by -
! physics tests-and control rod exercises, . partial movement of all-rods every
- the shutdown rode shall be fully with-j.
two weeks.- Rods which.have been exercised within the '
i ..
.I l ..
Amendment Nos. 62 & 59 40 f ;
i -_______-_ - _ - - _ _ _ _ _ - _ _ _ _ . .___ _ : _ - _ _ _ __ -_ _ _ _ _ _ _ _ - - - _ - - _ _ _ _ . _ _ - _ - _ - _ _ _ _ _ _ . . _ _ _ . _ _ _ _ - _ _ - _ _ _ - _ _
-- n . - -
^-
-- -. .__ ~ .. -.
~
~ _ ,. . ...
^ '
LII'.ITING COI;DITION F0il OPERATI0tl SURVEILLANCE REQUIREMENT drawn and the control group rods 4.2.1.D.l. past two weeks during normal 3.2.1.D.I.
shall be no further inserted than operation need not be verified the limits shown on Figure 3.2-2 Control rod bank positions with for Unit I and Figure 3.2-4 for respect to its insertion limit Unit 2 for 4-loop operation and shall be verified once per Figure 1.2-3 for Unit I and shift.
Figure 3.2-5 for Unit 2 for 3-loop ,
operation.
1 Control bank insertion may be 2. Control rod banh; worths shall
- 2. '
be measured following each further restricted if the measured control rod worth of all rods, refueling outage.
less the worth of the most reactive .
rod (worst case stuck rod), is less
' than the reactivity required to provide the design value of avail-able shutdown as shown in Figure 3.2-1.
- 3. During physics tests and cor. trol 3. Not applicable'.
rod exercises, the insertion limits need not be observed, but the lim-its in Figure 3.2-1 must be ob- .
" served except. during the low power:
physics test to determine total control rod worth and shutdown margin. For this t st the reactor may be critical with all. full s length control rods fully inserted, except for the predicted most reactive rod.
- 4. Three reactor coolant pumps per 4. prior to proceeding from hot '
unit shall be operating whenever shutdown to hot standby, verify a reactor is critical except that three reactor coolant pumps during natural circulation test, are operating except during (power <0% full power) or low natural circulation tests or power piiysics testing. Iow power physics testing.
4 Amenilment flos. 62 7. d) -
41
ii l.IF.ITiti3 COIIDITION FOR OPERATIO!! SURVEILLANCE REQUIREMENT (, s t
3.2.1.D 5. Reacter power shall not be increased 4.2.1.D 5. Not Applicable above 60% of rated power with only three reactor coolant pumps in
- operation unicss the overtemperature ' '
AT trip setpoint and the P-8 Inter-
- lock for three loop operation has been set in accordance with speci-fication 2.1.1.B.4.
1 * .
(
F. . Hod Bank Assignment E. Rod Bank Assignment
- Rod Bank Assignment shall be as Rod Bank Assig'nment shall be
- delineated in Figure 3.2-8. Except verified after each refueling during physics tests, the sequence of outage, for the refueled unit.
I withdrawal of the control banks, when going from zero to 1001 power, is A, B, C, D with control bank overlap.
F. Boric Acid System (per unit) F. Boric Acid System (per unit)
~
- 1. A reactor shal,1 not be taken from 1. Surveillance and testing of hat shutdown to hot .atandby unless the boric acid system shall l' the following conditions exist: be; performed as follows:
-' a. Boric: acid tank level,
- a. One boric acid tank for that ,
reactor contains at Icest 5140 concentration _and tempera-
' ture shall be verified gallons of 11.51 (but not great-er than 131) by weight boric prior to startup and acid solution at a temperature weekly thereafter.
a of at least 145'F. ,
i i -
i Anenduent-ilos. 62 7. 59 . 42
. - . . - , , . - - .. - . _ _ ~m - =_ _ - .m_ . _ . - _ m
4'
~ - .
)
), ) ,
~
LIF:ITItG CONDITION F0lt OPEHATION
' SURVEILLANCE'REGUIREMENT (
i Primary System noron Concentration 4.2.1 G. Primary System Boron Concentration
! 3.2.1 G. - Changes during Cold. Shutdown-Changes during Cold Shutdown When a boration or dilution oper- The opera' tion of at least one
- reactor coolant pump-or one ation is in progress, at least one l
reactor coolant pump or one resi- residual heat removal loop shall-
- he verified before the start of dual heat removal loop shall be
! - a boration or dilution operation.
operating.
{: .
l' ..
Reactivity Anomalics H. Reactivity Anomalies II .
^
A normalization'of the computed Reactivity anomaly evaluations shall' be performed .folloviing i boron concentration as a function. startups:after shutdowns of 72
- of'burnup shall be compared with tht hours .or- longer- duration but octual boron concentration of the shall not be required more than
' coolant. If the difference between the observed and predicted steady- once if more than one such shut-j down occurs in a two month, period. -
state concentrations reaches the .-
equivalant of one percent in ,
l reactivity, the NRC shall be noti- ~
ficd within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> and an evalu-1 ation as to the cause of.the dis-
{
crepancy shall be,made and' reported -
to the NRC within 30 days.
l i
,1 4
1 :
t I 44
! Amend:Ocht. flos. 62 & $9 - .
+
- - _ . . . . . . _ _ _ = - . _ . . _ . - _ _ _ . . . _ . _ _ - = _ _ . _ _ - - - .
_ _ _ . - - _ - _ - - _ _ - _ - - - _ . _ . - - - . - - _ _ ,n_ _
. n - - - - . ,a
I.II:I Tl HG CONDITION FOR OPERATION SURVEILLANCE REQUIRalENT 3.2.2.A 4.2.2.A 2.2.c (Cont i suied ) 2.2.c.
()) Power betueen the maximum and (2) A flux dif ference alarm shall minimum limits specified in indicate non-conformance with 3.2.2.A.2.2.a. the 13% AI target band for BASE LOAD operation. If the (2) AI uithin theA 1 target band alarm is temporarily out of as per section 3.2.2.A.4 and service, conformance with the 3.2.2.A.S. except use 13% d I applicable limit and the flux target band ins.taad of the +6, dif ference shall be logged
-7461 target band. hourly for the first 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />
. and half-hourly thereafter*
2.2.d if any of the requirements of Section 3.2.2.A.2.2.c. are not maintained then 2.Ld m WWh power must inunediately be reduced to below the power limited by APDMS typ 4.2.2.A.3 surveillance (Section 3.2.2.A.2.1.) and APD:ts type surveiilance must be The reference eauilibrium indicated initiated if the power is above P T- axial flux difference as a function -
of power level (called the target 3.2.2.A.3 The target flux dif ference at a given flux dif ference) shaU be determined at least once per equivalent full power level, po, is determined by power quarter. The target dif ference noting the indicated axial flux difference should be updated every effective at the power level with equilibrium full p wer m nth. This may be done xenon conditions established in the using the measured value for that core and with the full length rod bank l more than 190 steps withdrawn. P for m nth or by linear extrapolation the purpose of determining the taOget using the two most recent measured
' value, should be as high a power level values. The initial tarcet flux dif ference on a reload may be as practicable. The target flux dif-determined from design predictions.
ference at any other level, P, is caual .
to the target value of PO multipled by the ratio, P/PO. - 4 6b -
Amendment Hos. 62 & 59
m t i
~
- lillVI'.II.I.Afl('E IlF.OllIlll'rIF.!!T j .IIliTIIlil a'OllDITIOli l'Olt Ol*l
- ltNt'lOl!
t U. For the pu rpmse of determining 4.2.2.A n. tbt. agpl icalst e.
3.2.2.A
. penalties associated with deviations from the target band, t ime for use in applying Items 6.1 q
and 7.2 above shall be accumulated j in the following manner: .
4
. 8. l' For deviations at or below i 50% power, time shall be i accumulated auch that a I 1 minute actual deviation f equals a 1/2 minute accumulative penalty in
! applying Items 6.1 and 7.2
! above.
l l 8.2 For deviations above 50%
^
power, time shall be ,
a accumulated in a 1 fyr"1
- time basis in applying Items.
l 6.1 and 7.2 above.
4 1
4 i
a i
i 47D Amendnient flos. 62 & 59 i
( J LIMITING CONDITION FOR OPERATION SURVEILLANCE REQUIREMENT 3.2.2 B. Quadrant Power Tilt Ebtio Limits 4.2.2' B. Quadrant Power TiltFbtio
- 1. If an indicated quadrant power tilt. 1. Quadrant power tilt ratio shall be
.' ratio exceeds 1.02, except for calculated and logged along with physics tests, then within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />, the individual' upper'and lower one of the following steps shall be excore calibrated outputs as taken: follows:
- a. Correct the tilt, or a. Once each shift at power levels greater than 50%.
- b. Determine by measurement the b. Four times a' shift and following core peaking factors and apply a load change of more than 10%
Specification 3.2.2.A, or power a,t any power level above
. 50% if one or both quadrant c; Restrict core power level so power tilt alcrms are inoperable, l
as not to exceed full rating less 3% for each percent of ' '
quadrant power tilt ratio .
beyond 1.0.
- 2. If an indicated quadrant power 2. Not Applicable '
tilt ratio exceeds 1.02 for a - - .
period of 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> without known ,
cause, or if sudden tilt reoccurs intermittently withoet known .
cause, the reactor shall be put
, in the Ilot Shutdown Condition ,
within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />. Ilowever, operation below 50% of rated powert for testing and/or correcting the tilt, ,
4 sbail be permitted.
- Amendment Nos. 62 & 59 1
E
- i. .
I SURVET.LLANCE REQUIRF".E*C LI! ITH:3 CCI!DITION FOR OPERATIO21 .
) _
\'
3.2.2.I 3. If,-an indicated quadrant pouer 4.2.2.0 3. Hot Applicablet
, tilt exceeds 1.09, except for i physics testing, the recctor shall l be put' in the Itot Shutdown Con-1 .dition; however, operation below I 50? of rated power, for testing .
and/cr correcting the tilt, shall' he permitted.
l k' C. Instrumentation (per unit) C. ; Instrumentation (per' unit) l 1. Excore axial imbalance detector 1. Excore axial imbalance '
system detector system l a. The excore axial inhalance a.' Not Applicable j detector, system shall be ,
i recalibrated at least every
{
three effective full power
.non th n. The calibration shall '-
i he caeched each effective full power month usina the IPCORS 4
SYS?3:a and recalibrated if the l difference is > 15 The nin- -
f- imum requf.renents per flun nar used for the recalibration are:
- 1. At least 16 different '
i thinble traces, and i.
1
- 2. At least 2 differ'enct
{. - thinble traces, per j~ .
quadrant.
, b. If requirement 3.2.2.C.l.a b. - Not Appli' cable l cannot be' net, then power shall l be limited to 90t'of rated
. power for 4 loop operation u
~
and 60% of rated power for z 3 loop operation.
' Amendment Nos. 62 & 59 - 49'
,. .-_ - _ _ _ _ _ . - - _ = = - ... -. ., , . - . . . _ - - - . . . - - . . - . -..
h SURVEILLANCE REQUIREMENT '(,
j 1.EITII;G CORDMION FOR OPERATION i __
4.2.2.C.2. . Inoperable Excore Detector 3.2.2.t 2. Inoperable Excore Detector Channel Channel If an excore detector channel If an excore detector channel is .is inoperable, quadrant powertilt inoperable, quadrant power tilt ratio
{ shall be determined by periodically g ratio shall be determined by monitoring at least four monitoring incore thermocouples.
thermocouples per quadrant once an hour and after any load change ,
- i greater than lot at any power - i
- level above 50%.
i j !
t t
NIS Detector Temperature Control 3. NIS Detector Temperature Control
, 3.
One of the two reactor cavity a. Reactor cavity ventilation a.
, fan operation shall be ventilation fans (Unit I' i 1RVO12-1A, IRVOl3-1B) or (Unit II verified once a shift.
2RVO12-2A, 2 RVOl3-2 B) shall be operating whenever Tavg is great-er than 145'F.
If this condition cannot be met, b. Not Applicable.
h.
the reactor shall be brought '
l 2
to the Hot Shutdown Condition immediately. g i
!t i
i i
1 5) >
i Amendment Nos. 62 & 59
, . . , _ _ _ _ _ ._, _ _ _ . . ~ __ _.
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ii n
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a a e n l en d c c d c E t n aca i i o 1 C n q nn l p
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A i
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. L d n rda t t a t I o i an o' o r o -
E ' f
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M '
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n .
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. daef k td onrt r rnn2e'. t oii nyr r olit oa nn2 an -
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c se l o
( n l n. ic-n e en 1 l oo . il t r ddci t l ohn' l fl na lluvob roeel N n oi r rrn e r i i ae .
v 'c a y o r o uoo v O t thi tC f cht sd s o c' i ,I sti e n do es n I
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act l one n de son o l
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p I 2 p h i c pp n da n . pti m nw 2 .
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o L om1 nap i. ' 9
<1i nt td l eei o l go l arb3
. O n i n n ieS ce aubm i ncy 1
bco 5
F m e - i h n mp i t nq i nni e
1 d a er &
e m l l nt s e i r y ra iil l i are1 o rhml o N t n l a :t in ] eel t r s e1 l v uha n 2 O n I. u hw! a rp n ecnb e t aae n niiaA 6 I
T S y i fft o
, 'so R e eopr ef eehh . o r A.d t o e ohnr:
1 D cao Ro hrst y o3 yhi l nwae3 s I a a enl t l .l ct b i hp o D n d t rel tddp e 2 pi n a dco2 N
_ N o i f u oh o . . . f oonp h . ph e r noen .
O n 1 1 I omtf a h' c , I nraa T3awt e Armi1 t
n
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l d n u o
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n . d
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A I .
M C A I I
I.
I 3
2 3
, :;- ,4 ;?
O, O. ~
O LIlilTitid CONDITION FOR OPERATION l SURVEILLCCE REQUIREMENT -
3.2.3.D 2. Not more than one inoperable 4.2.3.B 2. Not Applicable.
control rod shall be permitted during power operation. If more than one rod is determined to be inoperable, the reactor ,
shall be placed in the hot shutdown condition within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.
- 3. 1f more than one control rod is 3. Not Applicable. .
- inoperable because of a Rod Urgent l Failure in the rod control system, the provisions of Specifications ,
3.2.3.B.1 and'3.2.3.D.2 above-shall not apply. If the affected i . assemblies cannot be returned to .
service within two hours, the reac-tor shall then be placed in the hot shutdown condition within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.
- 4. Deleted. 4. Deleted i l' .
l S. If an inoperable full-length rod 5. Not Applicable. . .
is located above the 200 step level and is capable of being tripped, '
- then the insertion limits in Piqure 3.2-2 for Unit I and Figure 3.2-4 for Unit II shall apply for 4 loop operation and the insertion limits in Figure 3.*2-3'for Unit I
.and Figure 3.2-5 for Unit II shall i apply-.for 3 loop operation.
i
~
t 2
Ang?ndment Nos. 62 & 59 ,
.tl' , , ,!i' 7 I I < _ ! ,'
. d n:' .
n gy e u alut lrv l o i' l aoi rp 1url a u e t
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ei r dcre, o o i o nn %
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cd i alt no h, s d
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t a
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l o dr os f
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1- nrbt lv ya up e2 i on ndt o o l oos p s ie oe l
l itlr fr l gt w d o th itf l
t olo o an .e ee .P it t ao il Q ihai st cinhs ce s ar n R_ d hr' dc iwoco nc d o, r I e nesp oa foiih ox o pe efo I
m oh . re iltht e R c poi ctst r claw. t t E
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dn oe .le l eoc pf i m o f so at . l e
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l i2 l j ,jjI ;j!' i!;jjj {jjijj4!!(j ;ilij,:I!!
- +
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4 LIMITIfiG CONDITION FOR OPERATION SURVEILLANCE REQUIREMENT 3- 3. 2. 3.:). ] 2. If the conili tionn of Secti on 4.2.3.n.1.a shall be chnrted .1 idi rectly"
- 3.2.3.D.1 cannot be met the I y cmore de t ec torn .ru'/or l reactor shall Im brnunht to thorrocounion and/or nove- ,
at least the !!ot Shutdown able incore de tectorn every condition within four hourn chift; or a r t.e r ' .'n v ror!
, and the reactor trip broal:- no t i on o f thn n on- i neli ca tin q l ern shal) rcrain open. rod, oxceedinn 12 n tepn ,
t
, which.'ve r occurn firnt.
- b. Durinq nne ra t ion bcInu 50?
n f ra teel pove r , no special' j 4. D'lli "a ra me te rn t'on i to ri n n in reitui red .
i A. Tbo following D!!R rela ted pa ra- 4.A l.
rach of the i>a raretern listed in metern nhall be main tained wi thin n,ecification 3.?.4.A nball be l the limits shown during opera ti on. verified to be within its limit at least once per 12 hourn.
- 1. Reactor Cnol an t syn ten Taun 4 t'Ot I R Tnop: 6566.3*P 2. The 'Teactor con]an t Syntem total
, . Tiliti:I: I.OnP : - jI flow ra to nbal] he determined to
- 2. Prnanu ri zer Prennure he within its linit by reasurement i Fotilt T/CP: 2 2220 pnia ,a t least once por 13 rann th n .
j (2205 pnio)*
TI! tnl: LOOP: P -
. 3. Reactor Coolan t System Tota l . .
Flow Rate ForlR LonP: ) 35t ,0fl0 ni"I
- T!!RI;C LOOP
- il 4
11 ifith'any of the above pa rane tern
, exceeding itn li bil t , re s tore the
- pa rame te r to within its limit with-in 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> or reduce t!.crmal power
- to less than 5?, of ra ted thornal power within the ne
- :t 4 hourn.
Limi t not applicable durinq either a thermal nower rarp increase in exce s or V rated thornal power per minute or a thermal power step increane la ercens of 10 ?. rated thernal nover.
? Parame ter limi ts for three loop operation ' tp he entablished prior to oneration above P-7 wi th
, l enn than r nur loops opera tino.
Anend'ient Nos. 6? 7. 59 5
4 4
2 '
j When an F gmeasurement is I.a k e n , both experimental For normal operation it is noF uccesseiry to
} ore ne and 'mann f act u ri n.1 i ol e ra nco mon t- ho allowed measure t-hone quanti t-ion cont i nnonal y. Tuntend 3 io. . . pi . . .i .e i a t . . . I I ow a n. . It h.c. 1.re n ih e i m t er .i I ha t ., t. ov I .I. .I ' ce s a d i n .- ;
- .. i .. . . . . a i in i h.
j t... a eniI . < . . . m.y. I al.. n uiih Ih. mov.ible i neos e- coiuli i in ui.. .o e ob:tei veil, i19 but ch a n no 1 - I .'n l o n .
- . .t....i... t ns mai .p i n.
- -: y . : cm a n.1 :In c.- p. cen t i r. Iimiin wiii h. met- tbene confitions as e .u:
j tin. appi ops i al .- allow.uu o los manul.uturin1 foilows:
t tolosance.
I 1. Control rods in a single bank move together j In the specil:ied l i m i t. of FNn , there is an B with no individuai rod insertion di f fering j ( .1 ) by more than 15 inches from the bank. demand
- pe rcen t. allowance for uncertainti.]s which means position. An indical_ed misalignment limit ~ '
- that normal ope ra t i on o f the core is expected to o f I, 12 steps, not including instrument j result in Fyn 6.1.55/i.0H. The logic behind the error, precludes a rod misalignment no. ,
larger uncertainty in this case is that (a) greater than 15 inches. With maximum '
abnormal perturbations in the radial power shape Instrumentation error considered the actual t' (e.g. rod misalignment) affectFyn,inmostcases rod misalignment is no more than 24 steps l wit.hout necessarily affect.ing Fg, (b) the operator or 15 inches.
i' has a direct influence on F throuqh movement of rods, and can limit it to tRe desired value, he 2. Control rod banks are sequences with over-
- has no direct cont.rol over F@n and (c) an error lapping banks as described'in Technical j in the tredictions i for radial power shape, which Specification 3.2.
may be detected during staitup physics tests can j be compensated for in F9 by tighter axial control, 3. The full length control ~ bank insertion limits are-not violated.
but compensation is less readily available, j
When a measurement of forPhn$nin F
error must be allowed for and 4 percent is the taken, experimental
- 4. Axial power distribution control procedures, appropriate allowance for a full core diap taken which are given in terms of flux differences with the movable incore det ector flux mapping., control or additional axial power monitoring system. and control bank insereion limits are
. observed. Flux difference refers to the tieasurements of the hot channel factors are required difference in signals between the top and as part of start-up physics tests and whenever bottom halves of two-section excore' neutron
! abnormal power distribution conditions requi re a detectors. The flux difference is a measure-reduct_i'on of core power to a level based on measured of'the axial offset which is defined as the hot channel factors. The incore map taken following difTerence in normalized power between the initial loading provides confirmation of the basic top and bottom halves of'the core.
nuclcar design bases.ineludi.ng proper loading The pcrmitted relaxaLion in F$n a1 lows radla1 i pa t.t erun. The periodie monthly incore mapp.ing provides additiosial annurance that the nuclear power shape changes with rod insertion limits, design bases remain inviolate and identify opera- It han been determined that provided the tional anomalies which'would, otherwise, affect above conditions 1 through.4 are observed, j these bases. these hot channel factorL1imits are met. .
- In Specifications-3.2.2, Fg is arbitrarily I limited for P.S 0.5.
Amendment Nos. 62 T, 59 .68
i .- ^
l
] .
i The trocedures for arlal power distribution conditions for measuring target flux dif ference every.
control referred to above are designed to month. For this reason, the specification provides-minimize the effect_s of xenon redistribu- t_wo methods for. upda ting the . target flux 'dif ference. .: ,
I tion on the axial-power distribution during The alarms provided are _ derived f rom' the plant pr o - -
l loadfollow maneuvers. nanically ccntrol cess computer which determines the one minute averages of flux difference in required to limit of the operabic excore detector outputs to monitor 1
! the difference between t.he current value 4I in:the reactor core-and alerts the operator when i of Plux pifference ( A I) and a reference AI alarm-conditions exist. Two types of alarm messa -
! value which corresponds to the full power ges are output. Above a: preset power _ level, an alarm -*
! equilibrium value of Axial offset (Axial message is output ~ immediately. upon , determining a
. of f set - AI/ fractional power) . The delta flux exceeding a preset band about' a target. ~4 L returence value of flux di f ference varies delta finx value. Below this. preset power-level, an ,
i vith power leveI and burnup but expressed alarm sessage is outputLif the AI excemled its allow- -
I 3 .u. . inial of fset it varies only with burnup. able limits for a preset cumulative amcunt of tine .
! in the.past.24' hours. For periods during which the
] The technical speci fications on power dis- alarm on flux.differente is inoperable,- manual sur ~ ,
j tritan _ ion control assure that.the FO limit veillance will be utilized to provide adequate warn-i is not exceeded and-xenon distributions ing of significant variations in expected flux _;'
- as e not developed which at a later time, differences. However every attempt should be made -
3
. would cause greater local power peaking to restore the alarm to an. operable condition as soon i even though the flux difference is then as ponnible. .Any deviations:from the t.arget band .
- within the. limits specified by the proce- durarsg manual logging shall be treated as deviations i dure. during the entire preceeding logging interval-and i approprate actions shall.be taken. This action is i
- necessary to satisty NRC requirements; however more' !
, The target (or reference ) value of flux frequent readings may be logged to minimize the-
! difference is determined as follows. At penalty associated with a deviation from the target *
} a ny t_ i me that. equilibrium xenon conditions band _ to justify continued operation at the.cnrrent have been established, the indicated flux ,
power.
i l difference is noted wit's the full length rod control rod bank more than 190 steps The times that deviations from the band occur are withdrawn (i.e. normal full power operatin9 normally accumulated by the computer.
- position appropriate for the time in life, ,
l- usually withdrawn farther as burnup proceeds).
- This value,. divided by the fraction of full l power at which the core was operating, is
! the full power value-af the target flux difference. Values for all other core power
} icvels are obtained by mult_lplying the full i power value by.the fractional power.. Since '
i the indicated equilibriutc value was noted, I
no allowances for excore detector' error are necessary and indicated deviation of the Al target' band are permitted from the indi-
- cated reference value. During periods where ext.ensive load following ~ is required, it may l be impractical to establish the required core
-6Da-.
! Amendment.Ilos. 62 & 59 i, ,
-a au_. s,__- . h va....-,.________,u-_ _ _ _ _-
_m -,____w wi_.2 , _ = _ _ . ._a--t- ..m _____ _ _ _ -a _ ,w _______________._________s_
_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _- w
_m-__. m e. m -_m___ --_ _ _ _ _ _ _ _ _ < _ . - . _ _ .-__
t Significantly'different f rora those resulting
, f rom operation within the target band. The i- instantaneous consenuences of being outside the ha :.d , provided rod insertion limits are observed, ,
is not worse tuan a 10 percent increment in -
ruiking factor tor . flux dif ference in the range-ii/ s 1) percent (+ / percent. to -[ percent in-3 i e.ited ) increasing by +1 percent for each 2
{ petcent d+ crease in rated power. Therefore,
- hile the deviation exists the pouer level is
- 1 i
- .iit.ml. to 90% of PT or lower depending on the i indicated flux dif ferenco '
j .
If, for any reason, flux di f ference is not coat.colted within t.he A I target-band for as long.
a period as one hour, then xenon distribution may i he significantly changed and operation at 50 percent is recuired to protect against potentially mre severe consecuences of so.me re cidents.
j 18 d iscu ssed above , the essence of the procediire
- is to maintain the xenon d istribution in the core n close t.o 't he eeeilibrium full power condition
, g c4 p-n s ibl e . This i.s accompli shed '.r; using the a as; n;st em to position the fell length control r+ : s t.o prodcce the recuired ind ica ted flux i d i f fc rarice.
.o: Condition II events, the core is protected f rom overpower and a minimum X.an of 1.30 by an nut.omatic protection system. Compliance with .
- operating procedures is assumed as e pre-candition for condition Il transients; however, cperat.or error and cauipment malfunctions are
- urately as=iumed to Icad to the cause of transients considered.
Anendment Hos. 62 & 59 - 69e -
1
I '
i In accordance with the approved westinghouse model DNB Parameters: The limits on the DNB re-as presented in WCAP 8381, no collapses are expected lated paranters assure that each of the throughout the fuel cycle of operation. The pre- parameters is maintained within the normal dicted minimum times for clad flattening are steady state envelope of operation assumed ruel peqion in the transient and accident analyses.
EFPH The limits are consistent with the initial FSAR assumptions and have been analytically 1 19,500 demonstrated adequate to maintain a minimuun 2 > 30,000 DMBR of 1.30 throughout each analysed 3* i 30,000 transient.
for Zion Unit 1. The predicted minimum times to The 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> periodic surveillance of these collapse for Unit 2 ares parameters thru instrument readout is aufficient to ensure that_the parameters ruel Region EPPH are restored within their limits following load changes and other expected transient 3 1 27,000 operation. The 18 month' periodic measurement of 2 > 30,000 of the RCS total flow rate is adequate to 3 > 30,000 detect flow degradation and ensure correlation of the flow indication channels with measured A design criterium requires that proposed reload flow such that the indicated percent flow fuel region exposure levels expected at the time will' provide sufficient verification of flow or discharge not exceed the predicted ministen rate on a 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> basis. ,
collapse tice. Cperation in the exposure range in i,-hich clad collapse is postulated is not per-A auadrant power tilt will be indicated ratted under these technical specifications.
by the excore detectors by the arrangement The prea seted minimum time to collapse for all of the current recorders on the control-reload fuel regions is greater than 30,000 effec- board. Pour 2-pen recorders are pro-tive full power hours. vided, the pens are grouped so that, in the absence of a quadrant power tilt, t_ne two
- Except that the four (4) Region 3 assebmlies to be used in the Extended Burnup Program for Yion Unit 2 have a predicted minimum time to collapse greater than 41000 EFPH.
Amendment Nos. 62 & 59
-ink traces coincide. Any divergence in power tilts. Analyses have shown that the traces indicate a power tilt. fractional increases in the x-y power .
Furthermore, a quadrant power tilt alarm peaking factor are less than or equal to is provided for the upper and lower sets twice the increase in the indicated quadrant of excore currents. power tilt ratio, i.e., an envslope with a 2:1 slope.
Quadrant power tilt ratio limits are based on As described above, ag uncertainty fa tor of the following considerations. Frequent power 1.08 is included in FEH and 1.05 in F . The re-tilts are not anticipated during normal opera- fore, a limiting power tilt ratio of 1.025 can be tion since this phenomenon is caused by some tole ated before the margin for uncertainty asymetric perturbation, e.g. rod misalignment, in F is depleted. However, a measurement x-y xenon transient, or inlet temperature mis- uncertainty is associated with the indicated match. A dropped or misaligned rod will easily ' quadrant power tilt ratio. Thus, allowing for a be detected by the Rod Position Indication System low measurement of power tilt ratio, the action .
or core instrumentation. A quadrant tilt by some level of indicated tilt ratio has been set at 1.02.
Other means (x-y xenon transient, etc.) would not An alarm is set to alert the operator to an appear instantaneously, but would build up over indicated tilt ratio of 1.02 or greater and that several hours and the quadrant tilt ratio limits action is required. To avoid unnecessary power are set to protect against this situation. They changes, the operator is allowed two hours in also serve as a backup protection against the which to Verify with in-core mappings and/or dropped or misaligned rod. (8) Operational exper- to determine and correct the cause of the tilt.
ience shows that normal power tilt ratios are less Should this actign not be taken, the margin for than 1.01. Thus, sufficient time is available to uncertainty in FQ is reinstated by reducing the recognize the presence of a tilt and correct the power by 2 percent for each percent of tilt ratio cause before a severe tilt could buildup. Durin9 above 1.0, in accord with the 2:1 slope envelope startup and power escalation, however, a large described above, or as required by the restriction tilt could be initiated. Therefore, the Technical on peaking factors.
Specification has been written so as to prevent escalation above 50 percent power if a large tilt The upper limit on the quadrant tilt ratio at is present. The numerical limits are set to be which hot shutdown is required has been set at commensurate with design and safety limits for 1.09 so as to provide protection against excessive DNB protection and linear heat generation rate as linear heat generation rate.
described below.
The quadrant power tilt ratio of 1.02 at which remedial and corrective action is required has been set so as to provide DNB and linear heat generation rate protection with x-y Amendment Nos. 62 & 59
The nuclear ion chambers located outside a In the event that an LVDT is not in service, '
reactor vessel measure the flux distribution the effects of a malpositioned control rod are of the top and bottom halves of a core. Core observable on nuclear and process information traverses in a few of the in-core instrument displayed in the control room and by core thermo-thimbles will establish that the excore flux couples and in-core movable detectors.
measurement equipment is properly calibrated.
Operating experience has established that the One inoperable control rod per unit is acceptable excore flux measurement system is of a reliable provided that the power distribution limits are design, and that the 10% load reduction, in the met, trip shutdown capability is available, and event of a recalibration delay, is an ultra provided the potential hypothetical ejection of conservative compensation. the inoperable rod is not worse than the case analyzed in the safety analysis report. The Operating experience at similar PWR plants has rod ejection accident for an isolated fully shown that quadrant power tilts determined by inserted rod will be worse if the residence time monitoring symmetric thermocouples are in very of the rod is long enough to cause significant good agreement with quadrant power tilts deter- non-uniform fuel depletion. The 3 day period mined from power distribution maps using the allowed for the analysis is short compared with Movable Detector System. the time interval required to achieve a signifi-cant non-uniform fuel depletion.
Operation of one reactor cavity vent fan per unit ensures an adequate flow rate of cooling The rod drop time of 1.8 seconds is based on the air to each NIS Detector (9). negative reactivity insertion rate used in accident analysis. (11)
The various control rod assemblies (shutdown banks, control banks A, B, C, D) (1) PSAR - Figure 3.2.1-8 are each to be moved as a bank, that is, (2) FSAR - Table 3.2.1-1 I
with all assemblies in the bank within one step (3) FSAR - Figure 3.2.1-11 (5/8 inch) of the bank position. Position (4) FFAR - Chapter 14 indication is provided by two methods: a digital (5) FSAR - Section 3.1.2 count of actuation pulses which shows the demand (6) FSAR - Section 3.1.3 position of the banks and a linear position in- (7) PSAR - Chapter 14, Appendix C dicator (LVDT) which indicates the actual rod (8) FSAR - Question 3.8 position (10). The rod position indicator (9) FSAR - Section 9.10.6 channel is sufficiently accurate to detect a (10) PSAR - Section 7.3 misaligned rod 15 inches away from the demand (11) PSAE - Figure 14-2 position of the bank. The indicated + 12 step (12) August 27, 1976 Order for Modification of permissible misalignment provides an enforceable License.
limit below which design distribution is not exceeded.
72 Amendment Nos. 62 & 59
I sij. ...
W
~
O, -
O "ph. .
LIMITING CONDITION FOR OPERATION - SURVEILLANCE REQUIREMENT 4.3.1.B.5 Reoorts
. A. Following each inservice insrec-i tion of steam generator tubes, the nunber of tubes plugged in o , each steam generator shall te reported to the Commission with-in 15 days.
. B. The complete results of the steam generator tube inservice inspection shall be included in the Special Report pursuant to Specification 6.6.3.(c).
The Special Report shall 1:e submitted 4
within 12 months following completion
- . . of the inspection. This report shall incl ude
5
~
1 Number and extent of tubes
! insvected.
l l 2. Location and percent of well
- thickness nenetration for each indicati,n of an imoer-i fection. ,
- 3. Identification of tubes nlugged,
~
i C. Results of steam generator tube
- inspections which . fall into cat-i egory C-3 and require crompt not-i ification of the Comm19sion shell
! be renorted pursuant to Soecifi-
! cation 6.6.2 prior to resumption ~
of plan't oceration. The written l follow up of this report shall i - provide a description of invest- -
l igation conducted to determine ,
! Amendment Nos. 62 & 59 74g -
i ^ ~ ' '
i _ _
._ mm ., - . _ . __.m. .. __ _. ._ ..m. , _ _ _ . _ . . .
4 i
t t
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cC .: L 03 - 0 000>0 t* a: CJ 3 tt . .
L *Z = 0 H 1 t.0. CO CL LCO 4 = 00tC
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- Loe O *: H : C 1 CCt A C c0cc4 hL *'.
. C
- LQOm* C
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I l.Iltl Ti tial t'OtililTint! IMil Ol' Ell ATinti 'illilVRTI T.. AtiCR llK8 IITill21EllT i _ __ _. ..__.__ ._ __ _ . _ . - - - , .. . . . _ _ _ _ _ _
4 t . l . .' It'on t- i nu. il) . 4.3.2
)' It , The 1 iini t Iinn . thown in l'igures 11 . Not Applicable 3,3.2-1 and 3. .t . 2 - 2 shall be re- ~
j calculateel periodically as required, .
, based on i.esults from the material -
] surveillance program.
j C. The secondary side of the steam C. Not Applicable generator must not he pressurized above
~
. ~
200 psig i f the f.emperature o f tire pri-mary and secondary coolant is below 700r D. Tile pressurizer heatup rate shall not D. Not Applicable exceed'100"I'/hr and the pressurizer l cooldown rate not exceed 200l'/hr.
i The spray shall not be used if the * '
.i temperature dif forence between the .
i pressurizer anil the spray fluid is .
greater than 320"I'. ,
f E. Ilydrostatic Testing -
E. Not Applicable i .
1 -
- 1. System inservice leal; and hydro- '
p tests shalI bu performed in .
, accordance with the requirements
- of'ASt1H Boiler and Pressure Vessel ,
- Code',Section XI, 1974 Edition, up- ~
1 to and including Summer 1975 Addendum. -
i.
i t
j .
?
l Amendment rios. 62 & 59 .
-..--y ,-.w-ye s-.,m,..,w., , - . .-,w ,3,,. -,w~, ,,,y - .m -
w -
e, , > .- - __s _ .. - - .- . -
vv--- 1yr . __2 _ ,
LillITING CONDITION FOR OPERATION SURVEILLA; ICE REGJIRICLJT 3.3.2.'F. Safety ' Injection Actuation 4.3.2 F. Not Applicable j
~
- 1. I f safety injection should occur .
- when a reactor is in the hot shut- ,
i down condition or above, the ,
4 reactor sha11 remain in the hot i shutdown condition until the status -
I of the reactor coolant sys tem integrity is determined. .
j 2. If the inspection and review' (Sec. 6.1.G.2.a(7) and , Sec. 6. 3) .
of ,the reactor coolant sys tem ,
integrity determines that:
t -
l a. -The injection did not affect reactor coolant system integ- , .
riLy the p1 ant may proceed-to power operation.
- b. The inicction did affcct ' ' ~
reac' tor coolant system
] i n teg ri t.y , the reactor _shall .
be placed in the cold shutdown -
condition within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. , ,
f
- 3. In.the event the ECCS is actuated -
4 and injects water into the Reactor Coolant System when Tavg 2350*F, a Special Report *shall be prepared _
and submitted'to the Commission within 90 days describing the .
circumstances of the actuation and the total accumulated actu'ation cycles to date. The current valve of the usage factor for each af fected safety .
' injection nozzle shall be provided in this Special Report whenever its valve exceeds 0.70.
4
(
~
Amendment flos. 62 & 59 'n j ,
) ) )
1,li<lTit:G COliD1 TION FOR OPERATION SURVEILLA!!CE REQUIREI4EiiT ,
- 1. 3 3. Leakage (per unit) 4.3 3. Leakage (per unit)
A. If the leakage rate, from other A. When Reactor Coolant System than controlled Icakage sources, pressure is greater than 500 such as the neactor Coolant pump psig, one of the following Control 1ed Leakage Seals, exceeds monitoring requirements shall 1 ypm and the source of the Icak- be performed (4. 3. 3. A.1 or age is not identified within twen- 4.3.3.A.2):
ty-four hours of detection, the reactor shall be brought to hot 1. Containme'nt activity shall shutdown within four hours. If l be continuously monitored the source of the leakage is not by radiation' detectors identified within an additional 24 RE-00llA or RE-0012A.
! hours, the reactor shall be brough' to a cold shutdown condition 2. Manual sampling of the con-within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. tainment atmosphere shall be performed once a shift.
D. If the sources of leakage are D. When Reactor Coolant System-identified and the results of the pressure is - greater th-in 500 -
evaluations are that continued psig at least'three of the j opera tion is safe, operation of following monitoring require-the reactor with a total leakage, ments shall be performed - (4.3.3.
l other than leakage from controlled B.1, 2, 3, 4, and 5):
l sources, not exceeding 10 gpm shall be permitted except as specified in 3.3.3.C below.
95 Amendment Nos. 67 & 59 l
! 3) ([) (-)
i 1,INiTIlid CO!!!)lTIOli FOR. OPEllATION ' SURVEILLANCE IlEQUIREMENT I
e 4 4
3, 3, 3, n 4.3.3.B 1. The' amount of Reactor Cooldnt System makeup
, water required to maintain
! pressurizer-level and j volume control tank level
- shall be recorded.
1 -
l- ,
]
- 3. Containment pressure, ,
j temperature and humidity
- shall be monitored.
- 4. The high temperature alarm l
(TE-401) in the reactor 4 head flange leakoff piping j shall be operable.
i t 5. The Reactor Vessel Leak -
~
Detection. system (RE-PR12A, j- , RE-PR12B, RY-PR12A, and j associated alarms)~shall i be operable.
C. If it is determined tliat. leakage- C. If the monitoring performed in i exists through a non-isolabic l sections 4.3.3.A and 4.3.3.B f ault which has developed in a indicates significant leakage ,.
Reactor Coolant System component a detailed ' investigation shall '
i body, pipe wall, vessel wall, or be performed to identify the
- pipe weld, the reactor shall be ,
sources and quantity of leakage.- '
j brought to a cold shutdown con-i dition within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> and corrective I action taken prior to resumption ,
of unit operation.
- I i .
j lAmendment flos. 62 & 59 , 96 ,'
__=_-_____ _ ___ ___________ - ________-_-____ - _ _ _ ___ _ _ _ _ _ _ _ - - _ _ - - -
i i
i t
i . - _. . _ . _ _ . . _ _ _ . _ . .., __ -. . . - - . .
4 1.2Ill T Itil; M >IIDI T I ofI la ilt Ol'l-;l:NI'l Oil SilllVI-Il I.I.AllCl; 111:00 i lll;fll;fJ l' .i I !
f 1
3.1. 4 4.3.4 D. itaterials Irradiat, ion Surveillance
, ' Specimen Inspection (per unit) 1
. Specimen. capsules to be used in the reactor vessel material surveillance .- ;
i ' program shall be withdrawn during the'- !
I ~ refueling period either immediately I i preceeding or following the Effective 'l~
i Full Power-Years (EPPY) of unit life -
l as foilows: ,
- Withdrawal '
- -Capsule Schedule l Designation (EFPY)-
T or U l.2 U or T 3.3.
X 5.8 Y 8.3 W,S,V,Z Standby ;
4 1 - ,
i '
t 1'
I l 1 i !
j ;
i, - 'i n
i r i
i I
}
f Amendment Nos. 62 & 59 10S >
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1
! s.7 dTEArt GErlEltATOlt EttEuGEt:CY llEAT Rf2tOVAL
~'
4.7 STEAM GEtIERATOP DIERGENCY llEAT RDtOVAL Apglicability: Applicability:
f Applien to auxiliary feedwater system and Applies to surveillance of auxiliary l steam generator safety valves, per unit. ~ feedwater system, and steam generator y safety valves per unit.
3 Obj ec t i_ve : Objective:
i l
To i nstere adesguate plant cooldown capabil- To insure av+11 ability of the above
- ity upon loss of normal feedwater flow and system and valves.
loss of main condenser vacuum. >
Speci f i ca tion: Specification:
- 1. Steam Line Safety Valves 1. Steam Line Safety-Valves l
l A. Twenty LSME code safety valves A.. Ten steam.ge'nerator safety valves .;
j (5 per steam generator) shall be per unit shall be tested for set 1
operable whenever the reactor is pressure at each refueling'oetage. -
4- heated above 350*F except as Testing shall be done by a -
specified in 3.7.1.C, 3.7.1.D, calibrated auxiliary lifting device i and 3.7.1.E. or by bench testing on compressed j gas. At least twc of the valves tested shall be from each orifice 2
. size ("Q" or "R"). All valves on a unit shall have been tested 4
at the end of each second refueling outage. The valves and the i corresponding ~ set pressures and orifice sizes are identified in Table 4.7-1. -
- n. Deleted n. Deleted i
i- !
l i
. 156 i /bnendment flos. 62 & 59 !
i
) )' )
SURVEILLAllCE REQUlltidlEllT 1.Itti T1 tlG COtiblTIOli FOlt OPERATIOli 4.0.4.B A flow balance test shall be per-formed on the affected lines during the next refueling outage following
! valve stroking or maintenance or other system modifications which might alter the E.C.C.S. flow characteristics.
E.C.C.S. flow rates for single l
pump operation shall meet the' following requirements under the minimum resistance configuration:
l 1. Charging pump cold leg injection plus seal injection shall not exceed 550 GPM;
- 2. SI pump hot or cold leg injection.
plus mini flow shall not exceed ,
650 GPM; e
- 3. The minimum charging pump cold leg injection through any 3 '
lines shall be 275 GPM; and l
1
- 4. The minimum SI pump cold leg injection through any 3 lines
] shall be 400 GPM.
2 1
Amendment. flos. 62 & 59 173^
i
< a
1 Component flame Component Number i- Itenidual lleat itemoval Pump-lh (2A) RIIO 01-1 A (2A)
Residual llent itemoval Pump-1B (2D) R11002-1B (28)
Residual lleat E:tchanger-1A (2A) ItII00 3-] A (2A)
Itesidual IIca t Exchanger-1B (2H) R11004-lO (2B)
MOV-SIB 81]A Recirculation Sump to RIIIt Pump Sinction valves MOV-SI8811B RtlST to RIIR Pump Suction Valves MOV-Bil8700A IIOV-RII8 700D l
- j. Isolation Valves from Reactor MOV-Ril870]
Coolant System to RilR Pumps MOV-RI'8 702 Residual lleat Ilcmoval Pumps ,
MOV-SI8812A Suction Valves ,f - MOV-SIH812B 4 #
4 l
1 i
ie Itesidual lleat Removal Pump System
' TABLE 4.8-3 i
Amendment flos. 62 & 59 187
. _ _ _ _ - _ - - _ _ _ . _ _ _ - _ _ _ _ _ - _ _ - - _ - _ - _ - _ . - _ _ _ _ - - _ _ _ _ _ _ _ - - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ - _ . . - _ _ . . - - . . _ _ _ - _ - _ _ _ _ _ - _ _ =_
4 All fuel tods are pressurized with helium leak rate test. The structure provides durims f abrication to reduce stresses and biological shielding for both normal and strains and to increase fatigue life. Design Dasis Accident situations. (1) l Fifty three full length rod clustet control ansemb1les consisting of 20 individual 11 0 % Ag - 151 In - St Cd alloy stainless 5.4.2 Containment System Structure steel clad rods are inserted into the guide thimliles at appropriate locations The lleactor Containment is in the shape of in the core. a cylinder with a shallow domed roof and a flat foundation slab. The cylin-Hurnable poison rods consisting of drical portion is prestre.ssed by a post-Isorosilicate glass sealed in stainless tensioning system consisting of horizontal s t.ce l tubes may be ised for reactivity and vertical tendons. The dome has a and/or power distritaution control, three-way post-tencioning system. The foundation slab is conventionally rein-S.4 Containment System forced with high-strength reinforcing steel.
~-
The entire structure is lined with one-quar-ter inch welded steel plate to provide vapor S.4.1 Design Dasis tightness.
The reactor containment completely The approximate dimensions of the Reactor encloses the entire Reactor Coolant Contai nment are: inside diameter, 140 System and assures that essentially feet; inside height, 212 feet; vertical no Icakage of radioactive materials to , wall thickness, 3-1/2 feet; dome thickness the environment would result even if 2'-8"; and the foundation slab thickness, gross failure of the Ileactor Coolant " feet. The containment encloses the System were to occur. The design of pressurized water reactor, steam generators, the containment liner with channels and reactor coolant loops and portions of the the penetrations permits a much more auxiliary systems and engineered safeguards sensi tive arsd accurate means of systems.
testing the containment leafage stat.us more frequently than is possible with a conventional integrated 297 Amendmen t flos . 62 f. '39 -
r s f yld o eed - g e orantf s i L r o ouan gdtt afpu hh n t c l d n t n a mh std a nili omI e . ewnaied e el f
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t p1 i .
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e.
W Durirn the reriods when the Supervisor cf the *
- l. The Supervisor of the Offsite Review and ,
Offsite Review and Investigative Fu"ctter is un-Investigative Function shall be appointed by the available, he shall designate this resTr~eibility Executive Vice-President (Construction, Production, to an established alternate who satisf'ies the formal and Engineering). The Audit Function shall be training and experience requirments fer the supervisor the responsibility of the Manager of Quality of the Offsite Review and Inve.;tigative :' unction.
Assurance and shall be independent of operations.
- a. Of fsite Review aM Investigative Function %e responsibilities of the personnel performing this function are stated below. %e Of fsite Review and The Supervisor of the Offsite Feview ard Investigative Function shall review:
Investigative function shall:
(i) provide directions for the review arv3 (1) %e safety evaluations for 1) chances to procedures, and investigative function and appoint a equi [ rent or systes as described in the safety senior participant to provide appropriete analysis report and 2) tests or exneriments direction, (ii) select each participant for cort:pleted under the provision of 10 GR Section this function, (iii) select a corrolment of 50.59 to verify that such actions did not constitute nnre than one participant who collectively an unreviewed safety question. Pro w sed channes possess background and qualifications in the to the Cuality Assurance Proaram description shall subject matter unler review to provide be revie4ed and aporoved by the Manaaer of Ouality ca prehensive interdisciplinary review coverage Assura ce.
under this function, (iv) ini pendently review and arprove the findings and reconvr.endations (2) o rooosed changes to prccedures, equi: rent or develorxxl by personnel perfor:rirq th -view systers which involve an unreviewed safety questien and investigative function, (v) ap and as defined in Section 50.59 10 CFR.
rernrt in a tirrely nanner all f kndt noncompliance with NRC requirement A /rovide (3) "ronosd tests or exnerirnents which involve an recommendations to the Station Superintendent, unreviewed safety question as defined in Section Division Vice President Nuclear Stations, Manager 50.59 10 CFR.
of Quality Assurance, Vice President (Nuclear (4)- Proposed changes in '"echnical Specifications or T Operations) and the Executive Vice-President operating licenses.
(Construction, ProductJon, and Engineering).
Amendment Nos. 62 & 59
4
~
Of fsit e Review and Investigative Function .
(Continued)
(5) Noncompliance with NRC requirements, Such responsibility is delegated to the'
- or of internal orocedures or Director of Ouality Assurance for instructions having nuclear safety Operating and to the Staff Assistant significance. to the Manager of Quality Assurance for maintenance quality assurance (6) Significant operating abnormalities or activities. ,
deviations from normal and expected -.
performance of plant equipment that Either shall approve the audit agenda and
- . affect nuclear safety as referred checklists, the findings and the report to it by the Onsite Review and of each audit. Audits shall be Investigative Punction, performed in accordance with the Company Quality Assurance Program and Procedures.
(7) Reportable Occurrences requiring 24 Audits shall be performed to assure hour notification to the Conmission. that safety-related functions are covered within a period of two years or
, (2) All recognized indications of an as designated below.
unanticipated deficiency in some aspect of design or operation of (1) Audit of the Conformance of facility
- safety related structures, systems operation to provisions contained or components. within the Technical Specification and applicable license conditions (91 Review and report findings and at least once per year.
recommendations regarding all 4
changes to the Generating Stations (2) Audit of the adherence to procedures, F Emergency Plan prior to icplemen- training and qualification of'the-tation of such change. station staff at least once per year.
(10) Review and report findings and recommen- (3) Audit of the results of actions dations regarding all itens referred by taken to correct deficiencies occuring j the Technical Staff Supervisor, Station in facility equipment, structures, systems or methods of ooeration that Superintendent, Division Vice President-
- l Suciear Stations and flanager of affect nuclear safety at least once
. Quality Assurance. per six conths.
- b. Audit Function (4) Audit of the performance of activities
, required by_the Quality Assurance .
+
The Audit function shall be the responsi- program to neet the Criteria of bility of the flanager of Quality Assurance Appendix "B", 10 CPR 50.
independent of the Production Departmeat.
(5) Audit of thE PacilityfEmergency Plan and implementing procedures.
. Amendment flos. 62 & 59 302 i
4 (6) Audit of the Facility Security Plan of Quality Aseurance or the Supervisor and implementing procedures. of the Offsite Reviev and Investigative i Function has the authority to order (7) Audit onsite and offsite reviews. unit shutdown or request any other action which he deems necessary to avoid (8) Audit the Facility Fire Prctection unsafe plant conditions.
Program and implementing procedures at least once oer 24 months. d. Records (9) An independent. fire protection and (1) Reviews, audits and recommendations loss prevention progran inspection and shall be documented and distributed audit shall be performed at least as covered in 6.1.G._1.a and 6.1.G.I.b.
! once per 12 months utilizing either qua?ified offsite licensee personnel - (2) Copies of documentation, reports, and or an outside fire protection firm. correspondence shall be keot on file at the station.
(10) An inspection and audit of the fire protection and loss prevention program
!- shall be performed by a qualified
. Outside fire consultant at least
< once per 36 months. ,
(11) Report all findings of noncompliance
- with NRC requirements and recommen-j- dations and results of each audit 1
to the Station Superintendent, the ,
Division Vice President-Nuclear Stations, j Manager of Quality Assurance, Vice j President (Nuclear Operations) Director of Nuclear Licensing, and to the Execu-I tive Vice President (Construction, Pro-duction, and Engineering).
- c. Authority
, g The Manager of Quality Assurance reports I to the Chairman and President and the i
Supervisor the Offsite Review and j Investigative Function reports to ' Director
- of Nuclear Safety who reports to the Chair-i man and President. Either the Manager ~
302A Amendment Nos. 62 & 59
,=,
~
- e. igocedures f. Personnel Hria ien administ rative procedures shalI be orepared and riai nt a i ned for t he of f-site (1) The persons, including consultants, reviews and i nves t_ iya t i ve functions nerforming the revieu and-investi-described i n speci fi ca tions 6.1.G.1. a. gative function, in addition to the These procedures shall cover the following: Supervisor of the Of fsite itevice 'and Investigative Function, hall i.' ave (1) Cont.cnt and method of snbmission of expertise in one or mere of the follow-uresentations to the Supervisor of the ing disciplines as appropriate' for the of f ;it e !teview and Investigative Function. subject or subjects being reviewed and investigated. ..!
(2) Use of corsni t toes and consul tants.
(a) nuclear power plant technology
( 3) Iteview and approval (b) reactor operations (c) utility operations - ,
(4) Detailed listing of items to be reviewed. (d) power piant de9ign - _
~i (e) reactor engineering ,
(S) Picthod of (a) apoointing personnel,' (fl radiological safety I (b) nerforming revieus, i nves t iga ti ons , (g) reactor safety analysis (c) 'rencrting findings and recommendations -(h) instrumentation and control of revievs and investigations, -(d) approving (i) metallurgy-reports, and (e) distributing reports. (i) any other aporo7riate (.isciplines-
_ required by unicus cnaracteristics (6) Det.crnining satisf actory completion ~ of of the facility. .
action required based on aooroved findings and recomendations reported by personnel (2) Individuals performing the Review and.
performing the review and investigative Investigative Function . shall possess a - z.
function. minimum fomal training,and-experience as listed below for u ch discipline.
(a) Nuclear Power Dlant Technology i Engineering _ graduate or .equiva- ,
lent with 5 years experience in the nuclear power field design and/or operation.
i Amendment.flos. 67 f. Sg -303 _.
.- . .x_ . . - , , . . . . _ . - .. ,
m
- p. l. Procedureu liar i t i ms identified in . C. Temporary changes'to procedures 6.2.A
. ;peci f i ca t i on 6.2. A and any clUmges
' and 6.2.B above may be made provided: ,
to stu-b ps ocedure:, sha1I lie aeviewe'l
.un t approveel by the Ope: alinq L. The intent of the original procedure I:nqineer and the Technical S t_a f f is not altered, Siipervisor in the areas of operation and fuel handling, and by the Maintenance 2. The change is approved by two Assistant Superintendent and Technical members of the plant management staff, Staff Supervisor in the areas of plant at least one of whom holds a Senior maintenance, j ust rument maintenance, Reacht)r Operator's bicense on the and plant inspection. Procedures for unit affected, items identified in spe ci fication 6 . 2 . 11 and any changes to such procedures 3. The change is documented, reviewed shalI be reviewed and approved by the by the Onsite Review and Investigative Technical Staff Supervisor and the function and approved by the Station Itad-Chem Supervisor. At least one Superintendent within 14 days of person approving each of the above implementation.
procedures shall hold a valid senior ope ra to r 's license. In addition, D. Drills of the emergency r.rocedures these procedores and changes thereto described in Specification 6.2.A.4 shall must have autherrization by the Station be conducted quarterly. These dri11s Superintendent before being implemented. will be planned. so that during the cocrse of the year, communication links
- 2. Work and instructions type procedures are tested ar.d outside agencies are which implement approved maintenance contacted.
or modification procedures shall be approved anil authorized by the .
.hiintetuince Assistant Superintendent where t he wri t ten aut.hority has been '
provided by LSc Station Superintendent.
The " Maintenance /llodification .'
Procedure" utilized for safety related work shall be so approved only if procedures referenced in the"!1aintenance/
!!odi fication Pa t.cedure" have been approved as required by 6.2.A.
Procedures which do not fall within the rerpiirements of 6.2. A or 6.2.11 may be aprroved by the Department Ileads.
Ai:iendnent flos. 67 1 59 308
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I 2
Ti!IS PAGE LEPT INTENTIONALLY DLANK 330 Amer:dment Nos. 62 & 59
5 4 '
^
Unit in Operating Mode (other than "
Position None 1.or 2 1& 2 cold shutdown.' i Shift Engineer or 1 1 2 Shift Foreman Nuclear Station Operator 1 2 3 Equipment Operator or ? 3 4 Equipment Attendant 1
l Pad-Chen Tecl*nician i l TOTAL 5 7 10 MINIMUM
- 5 6- 9
- The minimum number refers only for the case of shift shortage, caused by a sudden sickness or home emargency.
Notes:
- 1. SRO shall be present .on site at all times when there is fuel in the reactor.
- 2. A licensed man shall be in the control room at all times whenever fuel is in either reactor.
- 3. Two licensed men shall be in the control room during.
l reactor startups, shutdowns, operation, and other.
periods such as planned control rod manipulations.
- 4. For the period of Unit-1 and Urtit 2 Start up Test Program, two-licensed' men'per unit shall be in the ,
control room during any. operation of the reactor or plant which can cause changes in reactivity:which have not been verified previously by the_startup test program..
ZION S!!IFT MANNING CIIART -
Piqure 6.1,2 l
1 1
i Amendment-Nos. 62 & 59
l' ..
- a. .Onsite Review and Investigative Function (Continued) (5) Investigation of all noncompliance with -
.I requirements and shall prepare _ and forwar-
._ by personnel performing the Review and Investigative a report covering evaluation and recommendations. ,
Function; (v) report all findings of noncompliance to prevent recurrence to the Division Vice Presi-l with NRC requirements, and provide recommendations dent-Nuclear Stations and to the Suoervisor nf the
! to the Division Vice President-Nuclear Stations and the Offsite Review and Investigative Function.
I Supervisor of the Offsite Review and Investigative Function; and (vi) submit to the Offsite Review (6) Review of facility operations to detect and Investigative Function for concurrence in a potential safety hazards.
timely manner, those items described in Specifi-cation 6.1.G.l.a which have been approved by the (7) Performance of special reviews and investi-
- Onsite Review and Investigative Function. gations and reports thereon as requested by 1
the Supervisor of the Offsite Review and
. The responsibilities of the personnel perfonning Investigative Function.
i this function are stated below:
(8). Review of the Station Security Plan and (1) Review of: 1) procedures required by -shall submit recommended changes to the Specification 6.2 and changes thereto, Division Vice President-Nuclear Stations.
4
'2) any other proposed procedures or changes thereto as determined by the (9) Review of the Emergency Plan and station
- Plant Superintendent to affect nuclear implementing procedures and slull suteit 1
safety, recommended changes to the Division Vice Presi-dent-Nuclear Stations, j (2) Review of all proposed tests and
- . experiments that affect nuclear (10) Review of reportable occurrences and
- safety. actions taken to prevent recurrence.
- (3) Review of all proposed changes- to b. Authority the Technical Specifications, i The Technical Staff Supervisor is responsible 1 (4) Review of all proposed changes or to the Station Superintendent and shall make
, modifications to plant systems or recommendations in a timely manner in all areas
! equipment that affect nuclear safety, of review, investigation, and quality control
- phases of plant maintenance, operation and -
administr;tive procedures relating to facility 1
1 t
i
[
a Amendment Nos. 62 & 59 305
,- .- .~ _ ,, - . . . , ~. .
s.
I Authority (Continued) (1) Content and method of submission and 4 !
. b. presentation to the Station Superintendent, i 4
!. Division Vice Presider.t-Ncuear Stations and the i
' operations and shall have the authority to Supervisor of the Offsite Review and In-i request the action necessary to ensure compliance vestigative Function.
with rules, regulations, and procedures when in his opinion such action is necessary. The Station Superintendent shall follow such. (2) Use of consnittees.
) reconenendations or select a course of action (3) Review and approval.
that. is saore conservative regarding safe -
- . operation of the- facility. All such disagree- (4) Detailed listing of items to be reviewed.
4 ments shall be reported immediately to the ;
4 Division Vice President-Nuclear Stations and the (5) Frocedures for administration of.the quality
- l. - Supervisor of the Offsite Review and Investiga- control activities.
! tive Function, 1 i (6) Assignment of responsibilities.
- c. Records
- e. Personnel (1)' Reports, reviews, investigations, and ~
~
- reconenendations shall be documented . (1) The personnel performing the Onsite Review r
with copies to the Division Vice Presi- and Investigative Function, in addition to dent-Nuclear Stations, the Supervisor- of
- the Station Superintendent, shall consists the Offsite Review and Investigative of persons having expertise in
' Function, the Station Superintendent
.and the Manager of Quality Assurance. (a) nuclear power plant technology J (b) reactor operations
$. (2) Copies of all records and documentation (c) reactor engineering shall be kept on file at the station. (d) radiological safety and chemistry .
- d. -Procedures
.(e) instrumentation and control-i (f) mechanical. and electric systems.
Written administrative procedures shall be 'l (2) Personnel performing the Onsite Review '
l prepared and maintained for conduct of the and Investigative Fut.ction shall meet Onsite Review and Investigative Function. minimum acceptable levels as described l
j These procedures shall include the following: in ANSI N18.1.1971,: Sections'4.2 and 4.4.
i j
! .I i
i
- 1. .
Amendment Hos. 62 & 59 .306 i
l
. _ _ . - _ _ _ . - - _ _ _ ~. _ ~ . . - - - - - - . _ _ . _. _._
4 !
v'
- 2. Record of principal maintenance l 6.3 Action to be Taken in the Event of an Reportable activities, including inspection Occurrence in Plant Operation and repair, regarding principal
' items of equipment pertaining to j Any reportable occurrence shall be promptly nuclear safety.
- reported to the Division Vice President-Nuclear Stations or his designated alternate. Records and reports of reportable The incident shall be promptly reviewed 3.
and safety limit occurrences.
- pursuant to Specification 6.1.G 2.a(5) and a separate report for each reportable Records and periodic checks.
occurrence shall be prepared in accordance 4.
with the requirements of Specification inspection and/or calibrations l performed to verify the
- 6.6.B. Surveillance Requirements (See 6.4 Action to be Taken in the Event of a Safety Limit Section 4 of these Specifications) .
[
- is Exceeded are being met. All equipment , [
failing to meet surveillance -
a If a safety limit is exceeded, the reactor requirements and the corrective '
shall be shut down immediately and reactor action taken shall be recorded.
, operation shall not be resumed until authorized Records of changes made to the -
by the NRC. The conditions of shutdown shall 5.
- - be promptly reported to the Division Vice President- equipment or reviews of. tests and Huclear Stations or his designated alternate. experiments to comply with ;
l 10 CFR 50.59.
j The incident shall be reviewed pursuant '
to Specification 6.1.G.1.a and 6.1.G.2.a Records of radioactive shipments.
and a separate report for each occurrence 6.
shall be prepared in accordance with Records of physic tests and other Specification 6.6.B. 7. i i'
tests pertaining to nuclear safety.
J.5 Plant Operating Records
] 8. Records of changes to operating A. Record and/or logs relative to the- following procedures.
y 1
items shall be kept in a manner convenient '
for review and shall be retained for at 9. Shift Engineers Logs.
least five years.
j 10. By-product material inventory records -
j 1. Records of normal plant operation. and source leak test results.
including power levels and periods i of operation at each power level.
I t
314 Amendment Nos. 62 & 59 ,
. . , -_ _ , _ - .- - .,_ .